The present invention relates to nuclear boiling water reactors (BWRs) that utilize natural circulation and more particularly to improving the separation of steam in such reactor designs.
Existing large BWRs are of the forced-circulation type. In BWRs undergoing power generation operations, reactor coolant, initially in the form of sub-cooled liquid (e.g. water), is circulated by main coolant recirculation devices (e.g. jet pumps or mixed-flow motor-driven pumps) around a path a portion of which is comprised of the core lower plenum region (located at the bottommost section of the reactor), thence through the nuclear core and into a core upper plenum in communication with the core. Flow exiting the core upper plenum then passes through standpipes that lead to an assembly of steam separators. The reactor coolant exiting the nuclear core and passing into the core upper plenum is a two-phase mixture of steam and water, the proportion of which varies depending upon such factors as the power output from the fuel bundles, the amount of sub-cooling present in the coolant entering the fuel bundle, and the amount of flow through the bundles. This last factor depends on the power of the recirculation pumps and the hydrodynamic flow resistance presented by the fuel bundle geometry and wetted surfaces, and the amount of orificing representing restrictions to flow just prior to the coolant's entrance into the core fuel assembly.
Joining with the core effluent in the core upper plenum is the core "by-pass" flow, which is reactor coolant that has flowed from the core lower (entrance) plenum into the region external to the fuel assembly channels (but inside the core shroud), thence upwardly generally through the region occupied by cruciform-shaped control blades which stand in various degrees of insertion into the core, thence across the upper grid member (called the "top guide" ) which with its lattice-like configuration keeps the fuel assemblies in a regular array, and finally into the core upper plenum. This by-pass coolant stream at its discharge into the core upper plenum is comprised substantially of saturated liquid, with perhaps a small amount of steam. Within the core upper plenum, these two effluents--by-pass flow and fuel bundle exit flow--rapidly mix together and quickly lose identity from their origins.
Mechanical steam separation can be utilized to accomplish the separation of the stem from the steam/water mixture exiting the core. Some earlier BWR designs used free-surface steam separation where, just as in the household tea kettle, steam separates unaided from the free-surface, and saturated water remains in the bulk coolant, which in BWRs is recirculated back down the downcomer annulus. This type of steam separation is feasible so long as the steam-leaving velocity, i.e. the bulk average velocity of the steam taken across the available pathway flow area, is not large, i.e. is no greater than about 1.8 foot/second. If steam-leaving velocities exceed this value, there tends to be carried along with the steam an unacceptably high moisture content. The high moisture levels saturate the moisture-drying abilities of the steam dryer, thus resulting in an unacceptably high moisture content in the stem leaving the reactor and supplied to the turbine. When steam moisture contents are too high in the turbine steam flow, accelerated erosion can occur in the steam lines and on first-stage turbine blades and the efficiency of the turbine is reduced.
It is possible to obtain free-surface separation capabilities if the reactor pressure vessel (RPV) cross-sectional area is made sufficiently large. However, cost economies dictate that minimum diameter RPVs be used, so that mechanical steam separation has been developed to handle the high power output steam production levels of modern BWRs. In these latter designs, the steam bulk average velocity moving through the wet steam plenum region immediately downstream of the mechanical steam separators is about 5 feet/second.
The fuel assemblies grouped over the central region of the core tend to have higher exit steam qualities than do bundles located at the peripheral region of the core. It is desirable, nonetheless, that the flow rates and steam/water mixture proportions entering the steam separator standpipes be relatively uniform. To facilitate gaining more nearly uniform steam/water mixture for entry into the standpipes, the standpipe entrances are separated from the fuel assemblies by a distance of, for example, about 5 feet. Turbulent mixing occurring between the plumes leaving adjacent fuel assemblies, each with a different void content, is one mechanism acting to produce a more nearly uniform mixture which enters into the steam separator standpipes. More important to achieving flow mixture uniformity, however, is the hydrodynamic flow resistance represented by the standpipes, each with their end-mounted steam separators. Complete flow mixture uniformity entering the standpipes is at best difficult to achieve and, even with a five-foot separation between fuel assembly exits and standpipe entrances, it is not a design basis used for reactor performance evaluations.
The steam separator assembly consists of a domed or flat-heat base on top of which is welded an array of standpipes with a three-stage steam separator, for example, located at the top of each standpipe. One function of the standpipes is to provide a stand-off separation of the larger-diameter steam separators, which are generally arranged in a particularly tightly-compacted arrangement in which external diameters of adjacent separators are nearly touching with each other, so that separated liquid coolant discharged at the bottom of the separator has a more "open" flow path outwardly from the reactor longitudinal axis and out to the downcomer annulus region which lies at the inboard periphery to the RPV. A second purpose for the standpipes is a high-power-output natural-circulation reactor using mechanical steam separators is to provide juxtaposed regions which promote natural-circulation by means of a vertical region of two-phase (and, thus, low-density) coolant inside the standpipes which is juxtaposed against single-phase liquid coolant outside the standpipes in a so-called "downcomer region", in which region height provides a very significant part of the total natural circulation driving head for coolant flow circulation within the reactor.
The steam separator assembly rests on the top flange of the core shroud and forms the cover of the core discharge plenum ("core upper plenum") region. The seal between the separator assembly and core shroud flange is a metal-to-metal contact and does not require a gasket or other replacement sealing devices. The fixed axial flow type steam separators have no moving parts and are made of stainless steel, for example, to resist corrosion and erosion.
In each separator, the steam/water mixture rising through the standpipes (the "standpipe region") impinges upon vanes which give the mixture a spin, thus enabling a vortex wherein the centrifugal forces separate the water from the steam in each of three stages. Steam leaves the separator at the top of this assembly and passes into the wet steam plenum below the dryer. The separated water exits from the lower end of each stage of the separator and enters the pool (the "downcomer region") that surrounds the standpipes to join the downcomer flow. The steam exiting form all separators either may be in the same horizontal plane, or the separators may be arranged in a slightly crowned fashion at the center to compensate for the crowned water gradient of the pool surrounding the standpipes. See U.S. Pat. No. 3,902,876 and Wolf, et al., "Advances in Steam-Water Separators for Boiling Water Reactors", ASME Paper No. 73-WA/Pwr-4, Nov. 1973.
The steam separator assembly may be bolted to the core shroud flange by long hold-down bolts, or the separator together with the dryer assembly may be held down onto the core shroud flange by contact from the reactor head when the latter is assembled to the reactor vessel. The nominal volumetric envelope of the steam separator assembly is defined by the horizontal plane of its lower flange that contacts the core shroud flange, its cylindrical sides that provide part of the five-foot stand-off from the fuel assembly exits, the circumscribed diameter of the outermost row of standpipes, the circumscribed diameter of the outermost row of steam separators, and the generally horizontal plane of the exits to the steam separators.
The core upper plenum region in a BWR currently under design known as the "simplified boiling water reactor" (SBWR) is substantially devoid of other mechanical devices, pipes, or structures; whereas the core upper plenum of a BWR/6 and "advanced boiling water reactor" (ABWR) reactor design generally contains spargers and nozzles for core sprays, and distribution headers for core flooders, respectively. In both reactor types, these spargers/headers are located at the outer periphery of the core upper plenum, mounted below the core shroud flange so that the sparger/header is clear of the refueling removal path of peripheral fuel assemblies and, thus, do not become removed during core refueling operations.
With specific reference to a natural-circulation SBWR, it will be observed that there are no recirculation pumps to aid in coolant recirculation. Steam generation in the core produces a mixture of steam and water which, because of steam voids, is less dense than saturated or sub-cooled water. Thus, the boiling action in the core results in buoyancy forces which induce core coolant to rise upwardly, to be continuously replaced by non-voided coolant arriving from beneath the core in the core lower plenum region. As the coolant leaves the core, it rises through the core upper plenum region, then through the standpipes region, and finally into the steam separators. This voided mixture inside these standpipes continues to be less dense than non-voided coolant external tot he standpipes, resulting in the development of additional buoyancy force to further drive the coolant circulation. That this process is quite effective in promoting coolant recirculation can be noted from reported tests made in forced-circulation power reactors where the coolant circulation pumps are shut off. Even with their relatively short steam separator standpipes, reactor power levels of 25% and coolant flow rates of 35% of rated flow, are readily and safely maintainable.
The SBWR reactor is but modestly different form the forced-circulation BWR, with the most prominent differences being that the standpipes region is to be considerably longer in the SBWR (to develop a higher differential head), the core overall height may be somewhat shorter (for example, being 8 or 9 feet active fuel length versus 12.5 feet active fuel length in recent forced-circulation reactors), and the core power density will be somewhat lower. The severity of orificing--a means to promote hydrodynamic stability--at the entrance to the BWR fuel bundles may be lessened. The fuel bundle may have a larger diameter fuel rod in, for example, a 6.times.6 rod array, whereas the rod array for a forced-circulation reactor often is an 8.times.8 rod array. The design flow rates per fuel bundle, and the flow rates per steam separator, will be somewhat reduced in the SBWR design. Fuel exit steam quality will be approximately the same between the two designs. In the SBWR reactor design, no spargers or discharge headers are installed in the core upper plenum, while in the ABWR reactor, spargers or discharge headers are installed in the upper core plenum.
In some versions of SBWR reactors under study, the standpipes are very long while the core upper plenum is short. In other versions, the converse is true. The present invention is applicable equally in either version and to forced circulation BWRs.
The steam output from the nuclear boiler is coupled to a turbine generator which, in turn, is coupled electrically to the grid. A nuclear boiler pressure regulation control system is installed, the action of which changes the position of turbine steam control valves in such a way so as to maintain constant the nuclear boiler pressure as measured int he reactor steam dome.
Particular care must be taken to reduce all irreversible pressure drops in the design of natural circulation BWRs. This is especially necessary in the two-phase region where most of the resistance occurs. The steam separators in an SBWR are a prime candidate for improvement as they cause 1/3 to 1/2 of the irreversible losses in the natural circulation loop. The optimum design of a steam separator in a natural circulation reactor is different from that of a forced circulation reactor, as pressure drops should be minimized even at the cost of high carryover and carryunder values. Free surface separation removes all resistance and is marginally possible for the SBWR.
Preseparation of flow can be enhanced by having a transition piece between the standpipe and the core upper plenum. The flow regime in the standpipes is slug flow and approaches annular mist flow. Both flow types are already somewhat separated. Care must be taken in the design of steam separators that what is already separated is not remixed. There are several points in a steam separation design as set forth in U.S. Pat. No. 3,902,876, cited above, where mixing occurs. There is an area change between the standpipes and the separating barrels. The inlet swirler furthermore has a fairly large hub and the pick-off rings introduce eddys which cause not only irreversible pressure losses, but also mixing between the pre-separated phases.